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At least 775 records · Page 43

Production of Radioactive 22 Na in Core-collapse Supernovae: The Ne-E(L) Component in Presolar Grains and Its Possible Consequences on Supernova Observations

Presolar graphite grains carry the isotopic signatures of their parent stars. A significant fraction of presolar graphites show isotopic abundance anomalies relative to solar for elements such as O, Si, Mg, and Ca, which are compatible with nucleosynthesis in core-collapse supernovae (CCSNe). Therefore, they must have condensed from CCSN ejecta before the formation of the Sun. Their most puzzling abundance signature is the 22 Ne-enriched component Ne-E(L), interpreted as the effect of the radioactive decay of 22 Na (T 1/2 = 2.6 yr). Previous works have shown that if H is ingested into the He shell and not fully destroyed before the explosion, the CCSN shock in the He-shell material produces large amounts of 22 Na. Here we focus on such CCSN models, showing a radioactive 26 Al production compatible with grain measurements, and analyze the conditions of 22Na nucleosynthesis. In these models, 22 Na is mostly made in the He shell, with a total ejected mass varying between 2.6 × 10 −3 M ⊙ and 1.9 × 10 −6 M ⊙ . We show that such 22 Na may already impact the CCSN light curve 500 days after the explosion, and at later stages it can be the main source powering the CCSN light curve for up to a few years before 44 Ti decay becomes dominant. Based on the CCSN yields above, the 1274.53 keV γ-ray flux due to 22 Na decay could be observable for years after the first CCSN light is detected, depending on the distance. This makes CCSNe possible sites to detect a 22 Na γ-ray signature consistently with the Ne-E(L) component found in presolar graphites. Finally, we discuss the potential contribution from 22 Na decay to the Galactic positron annihilation rate.

79 ASTRONOMY AND ASTROPHYSICS↗

Facile Synthesis of Micrometer-sized Hierarchical Porous Si@C Anodes for High-Performance Lithium-Ion Batteries

Extensive studies have been conducted to determine if replacing graphite anodes with silicon is feasible since silicon possesses a high capacity and is an abundant resource; however, Si anodes suffer significant volume changes during cycling, which pulverizes its structure and subsequently causes rapid capacity fading. We used SiO2 nanoparticles (nSiO2) and low-cost coal-derived humic acid as a feedstock to synthesize high performance Si-C anodes for Li-ion Batteries through spray drying, calcination, and magnesiothermic reductions. Our Si-C anodes features micrometer-sized porous Si coated with a graphitized carbon shell (mpSi@C). The hierarchical graphitized carbon shell and porous silicon structure relieve the mechanical stress of the Si phase upon cycling, which stabilizes the structure. This mpSi@C composite design allows for a high initial discharge capacity of 2199.9 mAh g-1 at 0.1 A g-1 and a cycling performance of 68% capacity retention after 100 cycles at 1.0 A g-1. The multipoint contact between the Si anode and C structure allows for a remarkable performance rate of 566.3 mAh g-1 at 5.0 A g-1.

01 COAL, LIGNITE, AND PEAT↗

Silicon-silicon oxide-carbon composites for lithium battery electrodes and methods for forming the composites

Composite silicon based materials are described that are effective active materials for lithium ion batteries. The composite materials comprise processed, e.g., high energy mechanically milled, silicon suboxide and graphitic carbon in which at least a portion of the graphitic carbon is exfoliated into graphene sheets. The composite materials have a relatively large surface area, a high specific capacity against lithium, and good cycling with lithium metal oxide cathode materials. The composite materials can be effectively formed with a two step high energy mechanical milling process. In the first milling process, silicon suboxide can be milled to form processed silicon suboxide, which may or may not exhibit crystalline silicon x-ray diffraction. In the second milling step, the processed silicon suboxide is milled with graphitic carbon. Composite materials with a high specific capacity and good cycling can be obtained in particular with balancing of the processing conditions.

Anguchamy, Yogesh Kumar↗

ECAR-2322 As-Run Thermal Analysis Of The AGC-2 Experiment

The second Advanced Graphite Creep (AGC-2) experiment was designed to irradiate various types of graphite specimens at a temperature of 600°C. The specimens were irradiated in an instrumented leadout capsule experiment in the south flux trap of the ATR during cycles 149A, 1498, 1508, 151A, and 1518. Temperature was monitored using twelve thermocouples located at various elevations in the reactor core, and a helium-argon gas mixture was used for gas gap temperature control of the specimens. The purpose of this analysis is to calculate specimen temperature using measured data on reactor power and helium-argon gas flows, and as-run calculations of heating rates and displacement per atom (DPA) in graphite. The accuracy of the model is assessed by comparing measured and calculated thermocouple temperatures. Uncertainty in gas gaps may preclude an accurate temperature calculation. In these cases, adjustments are made to the thermal model in order to reconcile the measured and calculated thermocouple temperature and to ensure the accuracy of the calculated specimen temperature.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

ECAR-3386 AS-RUN THERMAL ANALYSIS OF THE AGC-3 EXPERIMENT

The third Advanced Graphite Creep (AGC-3) experiment was designed to irradiate various types of graphite specimens at a temperature of 900°C. The specimens were irradiated in an instrumented leadout capsule experiment in the east flux trap of the ATR during cycles 152B, 154B, 155A, and 155B. Temperature was monitored using twelve thermocouples located at various elevations in the reactor core, and a helium-argon gas mixture was used for gas gap temperature control of the specimens. The purpose of this analysis is to calculate specimen temperature using measured data on reactor power and helium-argon gas flows, and as-run calculations of heating rates and displacement per atom (DPA) in graphite. The accuracy of the model is assessed by comparing measured and calculated thermocouple temperatures. Uncertainty in gas gaps may preclude an accurate temperature calculation. In these cases, adjustments are made to the thermal model in order to reconcile the measured and calculated thermocouple temperature and to ensure the accuracy of the calculated specimen temperature.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Nanometer scale nonvolatile memory device and method for storing binary and quantum memory states

Example implementations include an electronic memory device with a metallic layer having a first planar crystalline structure, a first encapsulating layer including an encapsulating material having a second planar crystalline structure, and disposed adjacent to a first planar surface of the metallic layer, and a second encapsulating layer including the encapsulating material, and disposed adjacent to a second planar surface of the metallic layer. Example implementations also include a method of depositing graphite crystals onto a substrate to form a gate bottom layer, depositing BN crystals onto the graphite bottom layer to form a BN bottom layer, depositing tungsten ditelluride (WTe2) crystals onto the BN bottom layer to form a metallic layer, depositing the BN crystals onto the BN bottom layer and the metallic layer to form a BN top layer, and depositing the graphite crystals onto the BN top layer to form a gate top layer.

Xiao, Jun↗

Nanoscale Three-Dimensional Imaging of Degradation in Composite Si-Containing Anodes

The use of silicon (Si) in next-generation lithium-ion battery (LIB) anodes has the potential to dramatically improve electrochemical performance over current LIB graphite (Gr) anodes, due to silicon’s higher specific capacity.1 However, widespread implementation of Si-containing anodes is inhibited by issues such as significant Si volume expansion during lithiation and an unstable solid-electrolyte interphase (SEI), resulting in unreliable performance and poor cycle life. Currently, composite anodes with both Si and graphite active materials are used to increase capacity and mitigate some of the limitations associated with Si. In composite electrodes with a heterogeneous distribution of components with varying electrical properties (including Si, Gr, conductive carbon additive, and binder), it is important to understand the local distribution of each component to correlate with electrochemical processes, particularly localized degradation and heterogeneous aging, and to optimize performance. To investigate Si-containing composite anodes in the nanoscale, we use scanning spreading resistance microscopy (SSRM), a form of scanning probe microscopy (SPM) that probes local electronic resistivity. By examining the intrinsic electronic resistivity contrast between the anode components, separate phases can be distinguished and understood within the composite structure.2 This work studies the effect of electrochemical cycling in two different electrolytes on component distribution and aging by comparing the electrical and structural evolution of composite Si-graphite electrodes and SEI before and after charge-discharge cycling. 1. W. J. Zhang. A review of the electrochemical performance of alloy anodes for lithium-ion batteries J. Power Sources 196 13–24 (2011). 2. C. Stetson, Z. Huey, A. Downard, Z. Li, B. To, A. Zakutayev, C.-S. Jiang, M. Al-Jassim, D. Finegan, S.-D. Han and S. DeCaluwe: Three-Dimensional Mapping of Resistivity and Microstructure of Composite Electrodes for Lithium-Ion Batteries. ACS Nano Letters Accepted (2020).

ADVANCED PROPULSION SYSTEMS↗

Methods for forming silicon-silicon oxide-carbon composites for lithium ion cell electrodes

Composite silicon based materials are described that are effective active materials for lithium ion batteries. The composite materials comprise processed, e.g., high energy mechanically milled, silicon suboxide and graphitic carbon in which at least a portion of the graphitic carbon is exfoliated into graphene sheets. The composite materials have a relatively large surface area, a high specific capacity against lithium, and good cycling with lithium metal oxide cathode materials. The composite materials can be effectively formed with a two-step high energy mechanical milling process. In the first milling process, silicon suboxide can be milled to form processed silicon suboxide, which may or may not exhibit crystalline silicon x-ray diffraction. In the second milling step, the processed silicon suboxide is milled with graphitic carbon. Composite materials with a high specific capacity and good cycling can be obtained in particular with balancing of the processing conditions.

Anguchamy, Yogesh Kumar↗

AGC-2 Specimen Post Irradiation Data Package Report

This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume data necessary to derive density, elastic constants (Young’s modulus, shear modulus, and Poisson’s ratio) from ultrasonic time of flight velocity measurements, Young’s modulus from the fundamental frequency of vibration, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between pre- and post-irradiation properties is presented. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to irradiation, temperature, and applied load on specimens will be discussed in later AGC-2 post-irradiation examination analysis reports.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Carbide-based fuel assembly for thermal propulsion applications

Carbide-based fuel assembly includes outer structural member of ceramic matrix composite material (e.g., SiC—SiC composite), insulation layer of porous refractory ceramic material (e.g., zirconium carbide with open-cell foam structure or fibrous zirconium carbide), and interior structural member of refractory ceramic-graphite composite material (e.g., zirconium carbide-graphite or niobium carbide-graphite). Spacer structures between various layers provide a defined and controlled spacing relationship. A fuel element bundle positioned between support meshes includes a plurality of distributively arranged fuel elements or a solid, unitary fuel element with coolant channels, each having a fuel composition including high assay, low enriched uranium (HALEU). Fuel assemblies are distributively arranged in a moderator block and the upper end of the outer structural member is attached to a metallic inlet tube for hydrogen propellant and the lower end of the outer structural member is interfaced with a support plate, forming a nuclear thermal propulsion reactor.

Barringer, Eric A.↗

XE-100 modeling and simulation for neutronic analysis in MCNP6.2

XE-100 is a generation IV helium-cooled, graphite-moderated, pebble-bed reactor (HTGR). As part of the pathway toward a conceptually designing and licensing this reactor, an independent Monte Carlo model was created in MCNP6.2, and several distinct neutronic analyses were then performed. The double heterogeneity of TRISO fuel within graphite pebbles introduces unique modeling challenges related to particle and pebble clipping. The results show that for neutron and photon heating of ex-core components such as the reflector, RCSS, core barrel (CB), the model that contains clipping produces higher heating values. It is therefore concluded that removing clipping via compression of the particles and pebbles within the model distributes the neutrons and gammas preferentially toward the core center, and reduces the heating that is experienced toward the reactor periphery. Thus, the most conservative model for ex-core heating contains particle and pebble clipping. Also presented are results on the impact of chamfers that exist on the corners of graphite reflector blocks. As these chamfers could potentially create streaming paths, the neutron and gamma flux from the core to the CB were analyzed. It was determined that the chamfers do not significantly impact the neutron or gamma signatures on the CB, in that the shape of the neutron and photon flux on a detector imposed on the CB shows no preferential streaming path. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

"Source Term Modeling for Advanced Gas Micro-Reactors"

Maintaining the safety of the public, environment, and operating personnel is the most important factor in designing, operating, maintaining, and decommissioning nuclear reactors. In recent years, there has been a growing interest in the development of micro-reactors employing TRi-structural ISOtropic (TRISO)-coated particle fuel. In gas reactors, TRISO fuel plays an important role in the safety case for high temperature reactors because of the fission product retention properties of the fuel. This ability enables the use of a functional containment strategy for the reactor where multiple barriers are used to prevent fission product release to the environment. Part of the safety analysis of these advanced reactors is the assessment of radionuclide releases under normal and accident conditions through the multiple credited safety barriers. Using conservative assumptions, a mechanistic analysis can be performed to quantify these releases that combines the probabilistic assessment of failure with analytic solutions to radionuclide transport equations. Source term modeling for TRISO fuel has been performed for previous reactor designs; however, these models are outdated, in many cases proprietary, and need updates to be applied to the current state of TRISO fuel technology and alternative gas reactor core configurations [1]. Currently, the only publicly available source term assessment for gas reactors is an expert-based Monte Carlo simulation based on the effectiveness of the fuel kernel, coating layers, and graphite block in a modular high temperature gas reactor [2]. Thus, there is a need to develop a simple, versatile, and mechanistic model of fission product release and transport in gas reactor cores that could be applied to a variety of reactors through user inputs and reactor-specific radionuclide inventories. The release is calculated by the diffusion of the key safety important fission products through the kernel, silicon carbide (SiC), graphite for both intact and defective TRISO particles based on fuel and graphite temperatures in the reactor under normal operation. These releases from the fuel enter the coolant where they can plate-out on cooler surfaces. A clean-up model is included for designs with a coolant purification system to remove fission gases. This initial distribution of fission products in the reactor serves as an initial condition for potential releases under postulated accident conditions. The model then can calculate the fission product release for any transient temperature profile and fission product releases can then be used to assess radiological dose to the workers and the public using conventional dose tools. Data on the diffusion of fission products is based on historic German TRISO experiments and the more current Department of Energy (DOE) Advanced Gas Reactor (AGR) TRISO fuel development program. The model is coded in python with inputs and outputs in excel spreadsheets, as well as python plotting utilities to aid in the interpretation of the results. References: [1] INL, NGNP Mechanistic Source Term White Paper, INL-10-17997, July 2010. [2] David A. Petti, Richard R. Hobbins, Peter Lowry, Hans Gougar, “Representative Source Terms and The Influence of Reactor Attributes on Functional Containment in Modular High Temperature Gas-cooled Reactors,” Nuclear Technology, Vol. 184, p. 181-197, Nov. 2013.

07 ISOTOPE AND RADIATION SOURCES↗

AGR-5/6/7 CAPSULE 1 THERMAL MODEL WITH OFFSET GAS GAPS

Fuel compact temperatures are a crucial factor in assessing the irradiation performance of tristructural isotropic (TRISO) fuel particles. In the absence of direct measurement, fuel compact temperatures were calculated using a 3 D finite-element thermal model that was subject to simulation uncertainty. The most dominant factor in the uncertainty of calculated fuel temperatures is the gas gap uncertainty caused by the graphite holder nub-to-capsule clearance. This thermal model was revised to examine the most probable graphite offset position for Capsule 1 for four different days during the irradiation. In the analysis, the offset distance and azimuthal direction at both the top and bottom of the holder were varied. The best-fit offset was estimated based on the minimum root mean square error (RMSE) of the residuals (measured minus calculated thermocouple [TC] temperatures) for the operational TCs in Capsule 1. Near the end of the irradiation, when the graphite had shrunk and the gas gaps had grown, a peak temperature was found with maximum offset along the TC lines, with nickel introducing a potential mechanism for fuel particle failures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Evaluation of the Simplified Assessment Peak Equivalent Stress Design Limit Probability of Failure

The ASME Boiler and Pressure Vessel Code Section III Division 5 Subsection HH Subpart A (HHA) outlines requirements and qualification methods for graphite core components in high temperature reactor environments. One such structural qualification method is the Simplified Assessment outlined in HHA-3220, which represents a comparison of service level specific equivalent stresses to allowable stress values derived from target probability of failures (POFs). The target POF is based on the structural reliability class (SRC) for the components and service condition of interest, and is used to establish design limits for combined membrane and peak equivalent stresses. Combined membrane stress limits are derived from two parameter Weibull distribution fits of the graphite grade specific tensile strength data and the target POF. The combined membrane stress limits are then scaled based on the ratio of flexural strength to tensile strength (referred hereinafter as Rtf) to establish the corresponding peak equivalent stress limits. It is desired that the target POF is at least maintained for both combined membrane and peak equivalent design limits, especially with the use of Rtf. The assessment presented herein examines tensile and flexural data for medium and fine grade graphites relative to the calculated combined membrane and peak equivalent stress limits per HHA-3220 and illustrates the simplified assessment method results in anticipated, or at least conservative, peak equivalent design stress limits.

36 - MATERIALS SCIENCE↗

Identifying Challenges in Safeguards for Metallic Fuel Fabrication Facilities

As new advanced reactors gain popularity, there is an increasing interest in metallic fuel fabrication for fast reactors. While metallic fuels themselves are not a new idea, as many of the first reactors employed metallic fuels, new designs, compositions, and fabrication methods are appearing throughout the nuclear community. As the interest grows and facilities are constructed, both domestic and international safeguards will need to be heavily involved to support safeguards-by-design (SBD) measures from the start. This work compiles a review of historical and modern fuel types and fabrication methods, fabrication processes, safeguards gaps, and potential safeguards solutions. Metallic nuclear fuel types have been around for many decades and were included in some of the first reactors including the Experimental Breeder Reactor (EBR)-I and -II, the Fermi 1 reactor, the Integral Fast Reactor (IFR), and the Dounreay Fast Reactor (DFR). These reactors used various compositions including pure uranium (U) metal, U-zirconium (Zr) alloys, plutonium (Pu)-aluminum (Al) alloys, U-fissium (Fs) alloys, U-Pu-Zr alloys, and U-molybdenum (Mo) alloys [1, 2, 3, 4, 5]. These small alloying additions are included to improve the material properties of the pure U metal. The alpha-phase U (stable below 661C) suffers elongation in one direction causing grain boundary cracking and increasing creep rate due to irradiation growth, thermal cycling, and preferential crystal orientation. It is ideal to utilize the gamma-phase U (typically stable above 769C) by adding small amounts of alloying elements such as Zr or Mo to stabilize this phase down to room temperature [3]. Additionally, some research has been focused on U with transuranic (TRU) elements present, typically coming from the used fuel recycling process. Including these elements in fast reactor fuel can aid in the reduction of nuclear waste by burning minor long-lived actinides. However, the additions of TRU elements can cause concerns to arise when trying to fabrication or safeguard metallic fuels. A typical metallic fuel element is shown in Figure 1. Sodium is added into the cladding to create a thermal bond between the fuel slug and cladding wall. The fuel slug is then inserted and the end plug is welded on to the top of the fuel element. A gas plenum is left to create a headspace for gaseous fission products to escape rather than continue to build in the fuel itself [1, 5]. Other fuel element geometries exist as well, such as the Lightbridge twisted cruciform geometry shown in Figure 2 [6]. This design allows for better cooling performance and provides room for fuel rod swelling without impacting the fuel rod diameter. There are many different fabrication methods for metallic fuels, which is one of the many benefits of these fuel types. Many of these fabrication methods are relatively easy and cost-efficient. The most popular fabrication method is injection casting, sometimes called vacuum induction melting (VIM), shown in Figure 3 [4, 8, 9, 7, 10]. This method was largely used for EBR-II fuel fabrication. The injection casting system is contained inside of a vessel consisting of a Y2O3-coated graphite crucible surrounded by an induction coil with ZrO2-coated quartz molds suspended above the crucible. The fuel feedstock is placed inside of the graphite crucible and melted using the induction furnace. The induction furnace utilizes a dual frequency with the high frequency melting the feedstock and the low frequency causing stirring of the melted feedstock to form a homogeneous mixture. The mixture is heated to approximately 1600C in an argon environment. The vessel is evacuated and then the quartz molds are lowered into the graphite crucible containing the molten metal and the vessel is repressurized to inject the metal fuel upwards into the molds. The molds are removed and then shattered to release the fuel slugs. This fabrication method was used to fabricate 39,000 metallic fuel pins for EBR-II. While injection casting has been the most common metallic fuel fabrication method throughout the decades, many other methods have been explored including low-pressure gravity casting, microwave casting, continuous casting, centrifugal casting, coextrusion, and many others [11, 12, 8, 13, 14, 15]. Some of these methods aim to mitigate challenges that arise with americium (Am) volatilization during the casting process for TRU-containing fuel feedstocks, an issue with injection casting. Coextrusion is one of the methods explored at the Idaho National Laboratory (INL) and has been utilized for the initial fabrication tests of Lightbridge's unique fuels, as well as other metallic fuels with cladding coextruded. In this process, large billets are formed and machined and then inserted into a molten salt bath for approximately 30 minutes. The billets are then loaded into the extrusion press and extruded. This process can be seen in Figure 4 [15].

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Copper carbon composite wire with a uniform carbon dispersion made by friction extrusion

Copper-carbon composites are a group of materials with excellent mechanical, electrical, thermal, and tribological properties. However, bulk size copper-carbon composites made by the traditional manufacturing processes, like rolling or extrusion, fall short of reaching some of these properties predicted by theory or demonstrated only by samples at centimeter scale or smaller. The two main challenges to the successful scaling-up are: 1) to uniformly disperse carbon in the metal matrix; 2) to avoid degradation due to oxidation or reaction from overheating. In this work, we first demonstrate friction extrusion as a new method to make bulk-size void-free copper-carbon composite wires with homogenized carbon dispersion. Three different carbon varieties, graphite powder, graphene nanopowder, and carbon nanotubes, were added to the copper matrix with the concentration ranging from 0.5 wt% to 15 wt%. Special tooling, processing parameters, and procedures were developed, especially for high carbon content samples. Ten-fold reductions of both copper grain size and carbon particle size were achieved and attributed to the high shear deformation. Energy dispersive X-ray spectrometry indicates the carbon powder was refined to a sub-micron level and uniformly dispersed in the copper matrix. Compared with that of pure copper, the thermal capacity of the composite wire increases by 30 % while density reduces by 29 %.

Friction extrusion, Metal matrix composite, Graphe↗

Graphitizable Pitch from Microwave Plasma Pyrolysis of Natural Gas

An innovative approach of microwave plasma was utilized to convert natural gas into tar, from which a highly graphitizable pitch was derived using fractional distillation. The natural gas-derived pitch (NGDP) was thoroughly characterized, and the graphitizability of the carbonized NGDP was assessed using polarized light microscopy. The NGDP and, for comparison, needle coke, petroleum coke, and shot coke were subjected to graphitization heat treatment (GR) at 2500 °C. Results indicate that the graphitizability of the NGDP exceeds those of all industrial standard cokes. The GR-NGDP showed the highest degree of graphitization and crystallite size among all samples.

03 NATURAL GAS↗

Oxidation Activities

Slides for Oxidation Activities, oxidation rate, penetration/lathing, and strength after oxidation work. Also includes recent and current oxidation studies, background for rate behavior, strength behavior, density profile, and penetration measurements, oxidation resistant graphite coating development, and irradiated and unirradiated graphite oxidation rate behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗